This invention relates to a process for extracting and separating fission product molybdenum from an irradiated uranium-containing target material.
Technetium-99m is an extremely useful tool for medical applications and diagnosis. This radioisotope is used in a variety of applications in medical diagnosis. It is well suited for brain, thyroid, liver, lung blood pool and tumor scanning. It is preferred over the other radioisotopes because of the selective uptake by specific organs, its short half-life and low radiation dose rate which reduces the exposure of the patient to radiation. In addition, technetium-99m can also be used in industrial applications, such as measurement of flow rates, process control and the like.
Since the radioisotope technetium-99m has a short half-life (6 hours), it is common practice to use a molybdenum-99-technetium-99m generator to provide a supply of the technetium-99m. Basically, such a generator is made by sorbing the molybdenum-99 parent radioisotope, which has a 66-hour half-life, on an anion exchange material such as alumina. Subsequent decay of the molybdenum-99 produces the technetium-99m which can be selectively separated as needed from the generator by elution with a saline solution (which yields the technetium-99m as sodium pertechnetate).
Conventionally, technetium generators have been charged with molybdenum-99 which has been obtained by neutron bombardment of natural molybdenum or enriched molybdenum-98. A minor proportion of the molybdenum-98 is converted to its radioisotope molybdenum-99 by neutron capture. Radioactive molybdenum so prepared is referred to as "neutron product molybdenum".
Molybdenum-99 can also be obtained as a fission product from neutron bombardment of uranium-235. This "fission product molybdenum" is available in the form of sodium molybdate or ammonium molybdate solutions having a much higher specific activity than the maximum attainable from neutron product molybdenum. By using the more highly active fission product molybdenum for loading a technetium generator, a generator can be prepared which yields sodium pertechnetate eluates of exceptionally high radioactivity and which are consequently very desirable for certain medical applications.
The production of fission product molybdenum usually is done using a target comprising a mixture of aluminum and uranium, generally about 75 percent aluminum by weight with the balance being uranium highly enriched with uranium-235. This production requires an extraction process for extracting the molybdenum-99 from the target. The process currently used for extracting fission product molybdenum (molybdenum-99) from an irradiated aluminum-uranium target involves dissolving it in a strongly alkaline aqueous solution, such as sodium hydroxide. The resulting solution is filtered to separate the uranium and insoluble fission products from the molybdenum-bearing alkaline solution. The solution is acidified with sulfuric acid, and molybdenum-99 is separated from the solution by extraction with bis (2-ethyl-hexyl) phosphoric acid. This yields a molybdenum-99 containing extract from which residual soluble fission products are removed by stripping with organic solvents. This process is effective in extracting the molybdenum-99 from the irradiated target, but it produces a large volume of highly radioactive liquid waste. For example, approximately 33 liters of liquid waste is generated in the production of 4000 Ci of molybdenum-99. This presents a costly liquid waste disposal problem necessitating the use of difficult handling procedures. Accordingly, it is desirable to provide a process for extracting the molybdenum-99 from the target material and simultaneously separating the molybdenum-99 from the other fission products in a manner avoiding the creation of large volumes of liquid wastes.
Various processes have been devised for the separation of fission products from an irradiated target material. In Nuclear Science and Engineering, Volume 29, Number 2, August, 1967, at pages 159-164, A. W. Castleman, Jr. and I. N. Tang describe the results of an experimental study of the behavior of fission products released into helium and air from target materials of metallic uranium and a uranium-3.5% molybdenum alloy. This experiment used a thermochromatographic technique for investigating the nature of low-concentration gas species and this was applied to the release from these target materials of fission products barium, lanthanum, cerium, and molybdenum. Barium, lanthanum, and cerium were found not to be released into air in significant quantities because of the low volatility of their respective oxides, but molybdenum was released as MoO.sub.3 in air. The same authors also report the chemical nature of fission products iodine and cesium vaporized from irradiated specimens into helium and air in the Journal of Inorganic & Nuclear Chemistry, Volume 32, Number 4, April, 1970, at pages 1057-1064. The same approach was used by E. A. Aitken et al for an investigation described in an article in the Transactions of the American Nuclear Society, Volume 14, Number 1, pages 176-177. In this article, the reactions of fission products in urania-plutonia fuels held in stainless steel fuel cladding were investigated.
Kenji Motojima et al disclose a method of separating molybdenum-99 from irradiated uranium dioxide by sublimation in the International Journal of Applied Radiation and Isotopes, Volume 27, pages 495-498 (1976). The irradiated uranium dioxide is converted to U.sub.3 O.sub.8 by heating at about 500.degree. C. in an oxygen atmosphere and then the nuclides (Mo-99, Te-132, and Ru-103) are separated from the U.sub.3 O.sub.8 by heating at 1200.degree. C. under vacuum.
However it remains desirable to provide a process in which the separation of the molybdenum is conducted simultaneously with the oxidation of the target material.